ML19221B061

Srp,Revision 1 to Section 4.3, Nuclear Design

ML19221B061
Person / Time
Issue date:	03/31/1979
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-75-087, NUREG-75-087-04.3, NUREG-75-87, NUREG-75-87-4.3, SRP-04.03, SRP-4.03, NUDOCS 7907120335
Download: ML19221B061 (19)

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'S N['h'#)STANDARD REVIEW PLAN U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION

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SECTIO? 4.3 NUCLEAR CESIGN RFVIEW RESPONSIBILITIES Primary - Core Performance Branch (CPB)

Secondary - None i

I.

AREAS OF REVIEW The revi of the nuclear design of the fuel assemblies, control systems, and reactor core is carried out to aid in crnfirming that fuel design limits will not be exceeded during normal operation or anticipated operational transients, and that the ef fects of postulated reactivity accidents will not cause significant damage to the reictor coolant pressure boundary or impair the capability to cool the core and to assure conformance with the requirements of General Design Criteria 10, 11, 12, 13, 20, 25, 26, 27, and 28.

The review of the nuclear design under this SRP section, the review of the fuel system design under SRP Section 4.2, the reviet of the thermal and hydraulic design under SRP Section 4.4, a d the review of the transient and accident analyses under the SRP section for Lhapter 15 of the applicant's safety analysis report (SAk), are all necessary in order to confirm that the requirements defined above are met.

The specific areas of interest in the nuclear design include:

1.

Confirmation that design bases are established as required by the appropriate General Design Criteria.

l 2.

The areas concerning core power distribution. These are:

The presentation of expected or possible power distributions including normal l

a.

and extreme casas for steady state and allo-ed load-follow transients and covering a ful' range of reactor conditions of time in cycle, allowed control rod positions, a id possible fuel burnup distributions. The power distributiens should include power spikes from fuel densification.

b.

The presentation of the core power distributions as axial, radial, and local distributions and peaking factors to be used in the transient and accident l

analyses.

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The translation of the design power distributions into operating power distri-butions, including instrument-calculation correlations, operating procedures and measurements, and neces:,ary limits on these operations.

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Ihe requirements for instruments, the calibration and calculations involved in their use, and the uncertainties involved in translation of instrument readings into power o stributions.

Limits and setpoints for actions, alarms, or scram for the instrument systems e.

and demonstratico that these systems can maintain the reactor wi'Nin design power distribution limits f.

Measurements in previous reactors and critical experiments and their use in the uncertainty analyses, and measurements to be made on the reactor under review, including startup confirmatory tests and periodically required measurements.

g.

The translation of design limits, uncertainties, operating limits, instrument requirements, and setpoints into technical specificatiors.

3.

The areas concerning reactivity coefficients. These are:

The applirant's preser.tation of calculated nominal values for the reactivity a.

coefficients such as the moderator _oefficient, which involves primarily ef f ects f rom density char';es and takes the form of temperature, void, or density coefficients; the Doppler coefficient; and power coefficients. The range of reactor states to be covered includes the entire operating range from cold shutdown through full power, and the extremes reached in transient and accident analyses. It includes the extremes of time in cycle and an appropriate range of control rod insertions for the reactor states.

b.

The applicant's presentation of uncertainty analyses for nominal values, including the magnitude of the uncertainty and the justification of the magni-tude by examination of the accuracy of the methods used in calculations (SAR Section 4.3.3), and comparison where possible with reactor experiments.

c.

The applicant's combinati'>n of nominal values and uncertainties to provide suitably conservative values for use in reactor steady state analysis (primarily control requirements, SAR Section 4.3.2.4), stability analyses (SAR Section 4.3.2.8), and the transient and accident analyses presented

~n SAR Chapter 15.

4.

The areas concer ning reactivity control requirements and control provisions. These are.

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^ e (; sf ,.. +_' Q C' .,t. : '-: j~ j/ & 'f> g.O 'N W(y,)d$['"$)l$>4'W S / &,j'h 9 k IMAGE EVALUATION TEST TARGET (MT-3) 1.0 'P "5 E l' OH L E + il 2.0 l.l es }UE I l.25 V l.4 .l-Il== cl 1.6 d!. == 6" h^ k\\ M,y, 4 '% s"A e ' C, V~fD'* ~ . 9 ' l';, Q ,, g' .y g/,, cfy a. The control requir ements and provisions for control necessary to compensate for long-term reactivity changes of the core. These reactivity changes occur because of depletion of the fissile material in the fuel, depletion of burnable poison in some of the fuel rods, and buildup of fission products and transuranium isotopes. b. The sontrol requirements and provisions for control needed to compensate for the reactivity change caused by changing the temperature of the reactor from the hot, zero power condition to the cold shutdown condition. c. The control requirements and provisions for control needed to compensate for the reactivity effects caused by changing the reactor power level from full power to zero power. d. The control requirements and provisions for control needed to compensate for the effects on the power distribution of %e high cross section Xe-135 isotope. e. The adequacy of the control systems to as sure that the reactor can be returned to and maintained in the cold shutdown condition at any time during operati'n. f. The applicant's analysis and expetimental basis for determining tne reactivity worth of a " stuck" control rod of highest worth. g. The provision of two independent control systems. 5. The areas of control rod patterns and reactivity worths. These are: a. Descriptions and figures indicating the control rod patterns expected to be used throughout a fuel cycle. This includes operation of single rods or of groups or banks of rods, rod withdrawal order, and insertion limits as a function of power and core life, b. Descriptions of allowable deviations from the patterns indicated above, such as for misaligned rods, stuck rods, or rod positions used for spatial power shaping. c. Descriptions, tables, and figures of the maximum worths of individual rods or banks as a function of position for power and cycle life conditions appropriate to rod withdrawal transients and m d ejection or drop accidents. Descriptions and curves of maximum rates of reactivity increase associated with rod with-drawals, experimental confirmation of rod worths or other factors justifying the reactivity increase rates used in control rod accident analyses, and equipment, administrative procedures, and alarms which may be employed to restrict potential rod worths should be included. / 4.3-3 Rev. 1 d. Descriptions and graphs of scram reactivity as a function of time after scram initiation and other pertinent parameters, including methods for calculating the scram teactivity. 6. The area of criticality of fuel assemblies. Discussions and tables giving values of K f r single assemblies and groups of adjacent fuel assemblies up to the eff number required for criticality, assuming the assemblies are dry and also imm(rsed in water, are reviewed. 7. The areas concerning analytical methods. These are: a. Description.. of the analytical methods used in the nuclear design, including those for predicting criticality, reactiv

• .y coefficients, burnup, and stability.

i b. The data base used for neutron cro;s s(ctions and other nuclear parameters, Verification of the analytical methods by comparison with measured data. l c. 8. The areas concerning pressure vessel irradiation. These are: a. Neutron flux spectrum above 1 MeV in the core, at the co e bounJaries, and at the inside pressure vessel wall. b. Assumptions used in the calculations; these include the power level, the use factor, the type of fuel cycle considered, and the design life of the vessel, Computer codes used in the analysis. c. d. The data base for fast neutron cross sections. The geometric modeling of the reactor, support barrel, water annulus, and e. pressure vessel. f. Uncertainties in the calculation. The AB reviews the adequacy of !imits on power distrib tion during normal operation in connection with their analyses of the thermal-hydraulic e sign, under SRP Section 4.4. The ICSB, under SRP Section 7, reviews the adequacy of proposed instrumentation to meet the requirements for maintaining the r? actor operating state within defined limits. Rev. 1 4.3-4 147 002 II. ACCEPTANCE CRITERIA 1. The basic acceptance criteria in the area of nuclear design are the General Design Criteria (GDC) related to the re:ctor core and reactivity control systems (Ref. 1). CDC 10 requires that acceptable fuel design limits be specified that are not a. to be exceeded during normal operation, including the effects of anticipated operational occurrences. b. GDC 11 requires +r at the power operating range, the prompt inherent r.ut 7ar feedback chara-te i, tend to compensate for a rapid increase in reacti >ty. 3 c. GDC 12 requires that power oscillations which could result in conditions exceeding specified acceptable fuel design limits are not possible or can b r reliably and readily detected and suppressed. d. GDC 13 requires provision of instrumentation and controls to monitor variables and systems that can affect the fission process over anticipated ranges for normal operation, anticipated operational occurrences and accident conditions, mod to maintain them within prescribed operating ranges. CDC 20 equires automatic initiation of the reactivity control systems to e. assure that acceptable fuel design limits are not exceeded as a result of artit pated operational occurrences and tr assure automatic operation of syst ms and components important to safety under accident conditions. f. GC. 25 requires that no single malfunction of the reactivity control system (.his does not include rod ejection or dropout) causes violation of the accept-l able fuel design limits. g. GDC 26 requires tnat two independent reactivity control systems of different design be provided, and that each system have the capability to control the rate of reactivity changes resulting from planned, normal power changes. One of the systems must be capable of reliably controlling anticipated operational occurrences. In addition, one of tr.e.p tems must be capable of holding the reactor core subcritical under cold conditions. h. GCC 27 requires that the reactivity control systems have a combined capability, in conjunction with poison addition by the emergency ore cooling system, of reliably controlling reactivity changes under postulm ed accident conditions, with appropriate margin for stuck rods. i. GDC 28 requires that the effects of postulated reactivity accidents neither result in damage to the reactor coolant pressure boundary greater than limited local yielding, nor cause sufficient damage to impair significantly the capa-bility to cool the core. 10 003 4.3-5 Rev. 1 2. The following discus <'ons present less formal criteria and guidelines used in the review of the nuclear design. There are no direct or explicit criteria for the power densities and power a. distributions allowed during (and at the limits of) normal operation, either steady state or load-following. These limits are determined from an integrated consideration m el limits (SAR Section 4.2), thermal limits (SAR Section 4.4), iter 7) and transient and accident analyses (SAR [ scram limits o a Chapter 15). The design limits for power densities (and thus for peaking factors) during normal operation should be such that acceptable fuel design limits are not exceeded during anticipated transients and that other limits, such as the 2200 F peak cladding temperature allowed for loss-of-coolant accidents (LO.A), are not exceeded during design basis accidents. The limiting power distributions are then determined such that the limits on [ower dev ities and peaking factors can be maintained in operation. These limiting power l distributions may be maintained (i.e., not exceeded) administratively (i.e., not by automatic scrams), provided a suitable demonstration is made that sufficient, properly translated information and alarms are available from the reactor instrumentation to keep the cperator informed. The acceptance criteria in the area of power distribution are that the informa-tion presented should satisf actorily demonstrate that: (1) A reasonable probab' exists that the proposed design limits can be met within the txpected operational range of the reactor, taking into account the analytical methods and data for the design calculations; uncertainty analyses and experimental comparisons presented for the design calculations; the sufficiency of design cases calculated covering times in cycle, rod positions, load-follow transients, etc., and special problems such as power spikes due to densification, possible asymmetries, and misalinned rods. (2) A reasonable probability exists that in normal operation the design limits will not be exceeded, based on consireration of information received from the power distribution monitoring instrumentation; the processing of l that information, including calculations invo'.ved in the processing; the requiremer.ts for periodic check measurements; the accuracy of design calculations used in developing correlations when primary variables are not directly measured; the uncertainty analyses for the information and processing system; and the instrumentation alarms for the limits of normal operation (e.g., offset limits, control bank limits) and for abnormal situations (e.g., tilt alarms for control rod misalignment). Criteria for acceptaule values anc' uses of uncertainties in operation, instrumen-} tation numerical requirements, limit settings for alarms or scram, frequency Rev. 1 4 ~^ 6 147 004 and extent of power distribution measurements, and use of excore and incore instruments and related correlations and limits for cffsets and tilts, all vary with reactor type. They can be found in staff safety evaluation reports and in appropriate sections of the technical specifications and accompanying bases for reactors similar to the reactor under review (Ref. 2). The CPB has enunciated a Branch Technical Position CPB 4.3-1 for Westinghouse reactors l which employ constant axial offset control (Ref. 7). Acceptance criteria for power spike node's can be found in staff technical reports on fuel densification (Ref. 3). Generally, special or newly emphasized problems related to core power distribu-tions will not be a direct part of normal reviews but will be handled in special generic reviews. Fuel densification effects a"d the related power spiking and the ese of uncertainties in design limits are examples of these

dreas, b.

The only directly applicable GDC in the area of reactivity coefficients is GDC ll, which states " .the net effect of the prompt inherent nuclear feedback characteristics tend to compensate for a rapid increase in reactivity," and is considered tn be satisfied in light water reactors by the existence of the Doppler and negative power coefficients. There are no crit .t explicitly l establish acceptable ranges of coef.icient values or preclude the acceptability of a positive moderator temperature coefficient such as may exist in pressurized water reactor, at beginning of core life. The acceptability of the coefficients in a particular case is determined in the reviews of the analyses in which they are used, e.g., control requirement analyses, stability analyses, and transient and accident analyses. The use of l spatial effects such as weighting approximations as appropriate for individual transients are included in the analysis reviews. The judgeent to be made under this SRF section is whather the reactivity coefficients have been assigned l suitably conservative values by the applicant. The basis for that judgment includ?s the use to be made of a coefficient, i.e., the aialyses in which it is important; the state of the art for calculation of tre coefficient; the uncertainty associated with such calculations; experim<.ntal ch .s of the coefficient in operating reactors; and any required checks of tne coefficient in the startcp program of the reactor under review. c. Acceptance criteria relative to control rod patterns and reactivity worths include: (1) The predicted control rod worths and reactivity insertion rates must be reasonable bounds to.alues that may occur in the reactor. These values are used in the transient and accident analyses and judgment as to the 147 005 4.3-7 Rev. 1 adequacy of the uncertainty allowances are made in the review of the transient and accident analyses. (2) Equipme ,? rating limits, and procedures necessary to restrict potential rod w y insertion rates should be shown to be capable of performing the.. ions. It is a CPB position to require, where feasible, an alarm when any limit or restriction is violated or is about to be violated. d. There are no specific criteria that must be met by the analytical methods or data that are used by an applicant or reactor vendor. In general, the analytical methods and data base should be representative of the state of the art, and the experiments used to validate the analytical methods should be adequate and encompass a sufficient range III. REVIEW PROCEDURES The review procedures below apply in general to both the construction permit (CP) and operating license (0L) stage reviews. At the CP stage, parameter values and certain design aspects may be preliminary and subject to change. At the OL stage, final values of parameters should be used in the analyses presented in the SAR. The review of the nuclear design of a plant is based on the informM Ln provided by the applicant in the safety analysis report, as amended, and in n.eetings and discussions with the applicant and his contractors and consultants. Inis review in some cases will be supplemented by independent calculations performed by the staff or staff consultants. Files of audit calculations are maintained for reference by the reviewer (Ref. 5). l. The reviewer confirms, as part of the review of specific areas of the nuclear design outlined below, that the desiga bases, design features, and design limits 6re established in conformance with the GDCs listed in Section II. 2. The reviewer examines the information presented in the SAR to determine that the core power distributions for the reactor can reasonably be expected to f all within the design limits throu;hout all normal (steady state and load-follow) operations, and that the instrument systems employed, along with the information processing systems and alarms, will reasonably assure the maintenance of the distributions within these limits for normal operation. For a normal review, many areas related to core power distribution will have been examined in generic reviews or earlier reviews of reactors with generally similar core characteristics and instrument systems. A large part of the review on a particular case may then involve comparisons with information from previous applica-tion reviews. The comparisons may involve the snapes and peaking factors of normal and limiting distributions over the range of operating states of the reactor, the effects of power spikes from densification, assigned uncertainties and their use, calculation methods and data used, correlations used in control processes, instrumen-tation i+quirements, information processing methods including computer use, setpoints Rev. 1 4.3-8 147 006 for operational limits and alarm limits, and alarm limits for abnormalities such as flux asymmetries. 4 An important part of this review, at the OL stage, covers the relevant sections of the proposed technical specifications, where power distributions and related controls such as control rod limits are discussed. Here the instrument requirements, limit settings, and measurement frequencies and requirements are set forth in full detail. The comparison of technical specifications should reveal any differences between essentially identical reactors or any lack of dif ference between reactorr with changed core characteristics. Where these occur, the reviewer must assess the significance and validity of the dif ferences or lack of dif ferences. This review and comparison may be supplemented with examinations of related topical reports from reactor vendors, generic studies by staff consultants, and startup r epor ts from operating reactors which contain information on measured power distribu-tions (Ref. 4). l 3. The re ; ewer dete. mines from the applicant's presentations that suitably conservative reactiv ty coefficients have been developed for use 'n reactor analyses such as those for control requirements, stability, and transients and accidents. The l reviewer examines: a. The applit 5ility and accuracy of methods used for calculations including the use of more accurate check calculations such a; the use of Monte Carlo techniques for Doppler models. b. The models involved in the calculation such as the model used for effective ruel temperature in Doppler coefficient analyses. The reactor state conditions assumed in determining values of the coefficientt c. For example, the pressurized water reactor (PWR) moderator temperature coef fi-cient to be used in the steam line break analysis is usually based on the reactor condition at end of life with all control rods inserted except the most reactive rod, and the moderator temperature in the hot standby range. d. The applicability and accuracy of experimental data from critical experiments and operating rmactors used to determine or justify uncertainty allowances. Measurements during startup and during the cycle of moderator temperature coefficients and full power Doppler coefficients in the case of PWRs, and I results of measurements of transients during startup in the case of boiling water reactors (BWPs), should be examined. As part of the review, comparisons I are made between the values and uncertainty allowances for reactivity coeffi-cients for the reactor under review and those for similar reactors previously reviewed and approved. Generally, many essential areas will have been covered during earlier review; of similar reactors. The reviewer noses any lifferences 147 007 4.3-9 in results for essentially identical reactors and any lack of differences for reacte., with changed core characteristics, and judges the significarce and validity of any dif ferences or lack of dif ferences. I 4. The review procedures in the area of reactivity control requirements and control provisions are as follows: a. The reviewer determines that two independent reactivity control systems of different design are provided. b. The reviewer examines the tabulation of control requirements, the associated mcertainties, and the capability of the control systems, and determines by inspection and study of the analyses and experimental data that the values are realistic and conservative. c. The reviewer determines that one of the control systems is capable of returning the reactor to the cold shutdown condition and maintaining it in this condition, at any time in the cycl?. It is necessary that proper allowance be made for all of the mechanisms t1at change the reactivity of the core as the reactor is 'sken from the cold shutdow. state to the hot, full power operating state. The wiewer should determine that proper allowance is made for the decrease in fuel temperature, moderator temperature, and the loss of voids (in BWRs) as the l reactor goes from the power operating range to cold shutdown. d. The reviewer determines that one of the control systems is capable of rapidly returning the reactor to the hot standby (shutdown) conditior from any power level at any time in the cycle. This requirement is met by rapid insertion of control rods in all current light water reactors Proper allowance for the highest worth control rod being stuck in the ft 1-out position must be made. In PWRs, operational reactivity control is carried out by movement of control rods and by adjustments of the concentration of soluble poison in the coolant. The reviawer must pay particular at'.ention to the proposed rod insertion limits in the power operating range, to assure that the control rods are capable of rapidly reducing the power and maintaining the reactor in the hot standby condition. This is an important point because tne soluble poison concentration in the coolant could be decreased in order to raise reactor power, while the control rous were lef t inserted so far that in the event of a scram (rapid insertion of control rods), the available reactivity worth of the control rods on full insertirn would not be enough to shut the reactor down to the hot standby condition. e. The reviewer determines that each of the independent reactivity control systems is capable of controlling the reactivity changes resulting from planned, normal power operation. This determination is cade by comparing the rate of reactivity change esulting from planned, normal operation to the capabilities of each of Rev. 1 4.3-10 147 008 the two control systems. Sufficient margin must exist to allow for the uncer-tainties in the rate. 5. The aview procedures in the area of control rod patterns and reactivity worths are: a. The reviewer determines by inspection and study of the information described in sub;ection I.5 that the control rod and bank worths are reasonable. This l determination involves evaluation of the appropriateness of the analytical models used, the applicability of experimental data used to validate the models, and the applicability of generic positions or those established in previous reviews of similar reactors, b. The reviewer determinet the equipment, operating restrictions, and ao.ninistra-tive procedures that are required to restrict possible control red and bank worths, and the extent to which the alarm criterion in subsection II.2.c.(2) is ! satisfied. If the equipment involved is subject to frequent downtime, the reviewer must determine i f alternative measures should be provided or the extent of prcposed outage time is acceptable. c. The reviewer will employ the same procedures as in a, above, to evaluate the scram reactivity information described in subsection I.5. The scram reactivity l is a property of the reactor design and is not easily changed, but if restric-tions are necessary the procedures in b, above, can be followed as applicable. 6. The information presented on criticality of fuel assemblies is reviewed in the context of the applicant's physics calculations and the ability to calculate criti-cality of a small number of fuel assemblies. This information is rclated to informa-tion on fuel storage presented in SAR Section 9.1 and reviewed by the Auxiliary Systems Branch (ASB) under SRP Sections 9.1.1 and 9.l.2. The ASB reviewer assumes that the applicant's criticality calculations have been reviewi by CPB and are acceptable. 7. The reviewer exercisec vnfessinnal judgment and experience to ascertain the follow-ing about the applicant's analytical methods: The computer codes used in the nuclear design are described in sufficient a. detail to enable the reviewer to establish that the theoretical bases, assump-approximations for a given code reflect the current state tions, and numeric 3 1 of the art, b. The source of the neutron cross-sections used in fast and thermal spectrum calculations is described in suf ficient detail so that the reviewer can confirm 147 009 4.3-11 Rev. 1 that the cross sections are comparable to those in the current ENDF/B data files (Ref. 6). If modifications and normalization of the cross-section datu have been made, the bases used must be determined to be acceptable. c. The procedures used to generate,,roblem-dependent cross section sets are given in sufficient detail so that the reviewer can establish that they reflect the state of the art. The reviewer ;onfirms that the methods used for the follcwing calculations are of acceptable accuracy: the fast neutron spectrum calculation; the computation of the U-238 esonance integral and correlation with experimental data; the computation of resonance integrals for other isotoras as appropriate (for example, Pu-240); calculation of the Dancoff correction factor for a given fuel lattice; the thermal neutron spectrum calculation; the lattice cell calculations including fuel reds, control assemblies, lumped burnable poison rods, fuel assemblies, and groups of fuel assemblies; and calculations of fuel and burnable poison depletion and buildup of fission products and transuranium isotopes. d. The gross spatial flux calculations that are used in the nuclear design are discussed in sufficient detail so that the reviewer can confirm that the following items are adequate to produce results of acceptable accuracy; the method of calculation (e.g., diffusion theory, S transport theory, Monte n Carlo, synthesis); the number of energy groups used; the number of spatial dimensions (1, 2, or 3) used-, the number of spatial mesh intervals, when applicable; and the type of boundary conditions used, when applicable. e. The calculation of power oscilla' ons and stability indices for diametial xenon reactivity transients, ax .I xenon reactivity transients, other possible xenon reactivity transients, and non xenon-il.duced reactivity transients, are discussed in sufficient detail so that the reviewer can confirm for each item thatthemethodofcalculation(e.g.,nodalanalysis,diffusiontheory,transportI theory, synthesis) and the number of spatial dimensions used (1, 2, or 3) are acceptable. f. Verification of the data base, computer codes, and analysis procedures has been made by comparing calculated results with measurements obtained from critical experiments and operating reactors. The reviewer ascertains that the comparisons cover an adequate range for each item and that the conclusions of the applicant are acceptable. 8. The analysis of neutron irradiation of the reatc;r vessel may be used in two ways. It may provide the design basis for establishing the vessel material nil-ductility transition temperature as a function of the fluence, nvt. Or, it may provide the relative flux spectra at various positions between the pressure vessel and the reactor core so that the flux spectra for varion test specimens may be estimated. This information is used by the MIEB in determining the reactor vessel material l Rev. 1 4.3-12 surveillance program requirements and pre o ure-temperature limits for operation under SRP Sections 5.3.2 and 5.3.3. CPB reviews the calculational method, the l geometric modeling, and the uncertainties in the calculations under this SRP section. The review procedures for pressure vessel irradiation include determina-l tions that: a. The calculations were performed by higher order theory than di, fusion theory. b. The geometric modeling is detailed enough to properly estimate the relative flux spectra at various positions from the reactor core boundary to the pressure vessel wall. c. The peak vessel wall fluence for the design life of the plant is less than 20 2 10 n/cm for 'trons of energy greater than one MeV. If the peak fluence is found to be 3. eater than this value, the MTEB is notified. l IV. EVALUATION FINDING The reviewer verifies that sufficient information has been provided and his review supports the follc,w.ng type of evaluation finding, which is to be included in the staf f's safety evaluation report: "The applicant has described the computer programs and calculational techniques used to predict the nuclear characteristics of the reactor design and has provided examples to de.3onstrate the ability of these methods to predict experimental results. The staff concludes that the information presented adequately demonst ates the ability of these analyses to predict reactivity and physics character stics of the plant. "To allow for changes of reactivity due to reactor heatup, changes in operating conditions, fuel burnup, and fission product buildup, a significant amount of excess reactivity is designed into the core. The applicant has provided substantial information relating to core reactivity requiremeats for the first cycle and has shown that means have been incorporated into the design to control excess reactivity at all times. The applicant has shown that sufficient control rod worth is available to shut down the reactor with at least a %1k/k subcritical margin in the hot condition at any time during the cycle with the highest worth control rod stuck in j the 'ully withdrawn position. "On the basis of our review, we conclude that the applicant's assessment of reactivity control requirements over the first core cycle is suitably conservative, and that adequac' negative worth has been provided by the control system to assure shutdown capability. Reactivity control requirements will be reviewed for additional cycles as this information becomes available. We also conclude that nuclear design bases, features, and limits have been established in conformance with the requirements of General Design Criteria 10, 11, 12, 13, 20, 25, 26, 27, and 20. 147 Oli 4.3-13 Rev. 1 V. REFERENCES 1. 10 CFR Part 50, Appendix A, General Design Criterion 10, " Reactor Design"; Criterion ll, " Reactor Inhertnt Protection"; Criterion 12, " Suppression of Reactor Power Oscillations", Criterion 13, " Instrumentation and Control"; Criterion 20, " Protection System Functions"; Criterica 25, "Protectiori System Requirements for Reactivity Control Malfunctions"; Criterion 26, " Reactivity C~ntrol System Redun-dancy and Capability"; Criterion 27, " Combined Reactivity ' mtrol Systems Capability"; and Crit _rion 28, " Reactivity Limits." 2. St df safety evaluition reports and plant technical specifications. Examples of .t are: a. Safety Evaluatior Report, General Electric Standard Safety Analysis Report (GESSAR), Section 4.3, Docket No. STN 50-447, U. S. Atomic Energy Commission, November 1974. b. Safety Evaluation Report, Combustion Engineering Standard Safety Analysis Report (CESSAR), Section 4.3, Docket No. STN 50-470, U. S. Nuclear Pegulatory Commission, NUREG-75/ll2, December 1975. c. Safety Evaluation Report, Jamesport Nuclear Power Station Units 1 and 2, Section 4.3, Docket Nos. STN 50-516/517, U. S. Nuclear Regulatory Commission, NUREG-75/095, October 1975. I d. Safety Evaluation Report, Greenwood Energy Center Units 2 and 3, Section 4.3, Docket Nos. 50-452/453, U. S. Atomic Energy Commission, July 17, 1974. e. Technical Specificatior.3, Browns Ferry Nuclear Plant Unit 1 and Unit 2, Sections 2.1 and 3.2 through 3.5, License No. DPR-33 and 52, June 28, 1974. i. Technical Specifications, Millstone Point Nuclear Power Station Unit No. 2, Sections 2.1, 3.1, and 3.2, Docket No. 50-336, August 1, 1975. } g. Technical Specifications, D. C. Cook Nuclear Plant Unit 1, Sections 2.1 and 3.1 through 3.3, License No. DPR-58, October 25, 1974. h. Technical Specifications, Arkansas Nuclear One Unit 1, Sections 2.1, 3.1, and 3.5, License No. J.)R-51, May 21, 1974. 3. Staff technical reports on fuel densification: a. Regulatory Staff, " Technical Report on Densification of Light Water Reactor Fuels," U. S. Atomic Energy Commission, November 14, 1972. b. Regulatory Staff, " Technical Report on Densification of Babcock and Wilcox Reactor Fuels," U. S. Atomic Energy Commission, July 6, 1973. Rev. I 4.3-14 147 012 c. Regulatory Staff, " Technical Report on Densification of Exxon Nuclear B4R Fuels," U. 5. Atomic Energy Commission, September 3, 1973. d. Regulatory Staf f, " Technical Report on Densification of valf United Nuclear Fuels Corporation Fuels for Light Water Reactors," U.

5. Atomic Energy Commission, November <., 1973.

Regulatory Staff, "Technica, Report on Densification of Westinghouse PWR e. Fuel," U. S. Atomic Energy Commission, May 14, 1974. 4. Topical and startup test reports which are current and applicable to the reactor under review. Examples of these are: a. G. N. Kear and M. J. Ruderman, "A7 Analysis of Methods in Control Rod Theory l and Compari'an with Experimenu," GEAP-3937, Gene"al Electric Company, May 1962. b. J. S. Moore, " Power Distribution Control of Westiqqhouse PWR's," WCAP-7811, Westinghouse Electric Corporation, December 1971. c. J. O. Cermak, et al., " Pressurized Water Reactors pH-Reac;,ivity," WCAP-3696-8, Westinghouse Electric Corporation, October 1968. d. "Surry Power Station - Unit 2. Startup Test Report," Virginia Electric Power Company, July 31, 1973. e. J. E. Outz, " Plant Startup Test Report, H. B. Robinson Unic No. 2," WCAP-7844, Westinghouse Electric Corporation, January 19/2. f. R. H. Clark and T. G. Pitts, "^hysics Verification Experiments, Core I," BAW-TM-455, Babcock and Wilcox Company, June 1966. g. R.

4. Clark, " Physics Verification Experiments, Cores II and III," BAW-TM-458, Babcock and Wilcox Company, July 1966.

h. D. R. Jones and J. G. Harsum, " Field Testing Requirements for Fuel Curtains and Control Rods," NED0-10017, General Electric Company, June 1969. i. R. Bar ry, et al., " Nuclear Design of Westinghouse PWR's with Burnable Poison Rods," vCAP-9000-L, Revision 1, Westinghouse Electric Corporation, June 1969. j. G. V. Kumar, "Startup Test Results - Dresden NPS Unit 3," NEDC-10692, General Electric Company, December 1972. k. E. J. Deaa, " Quad Cities Units 1 and 2 - Startup Test Results," NEDC-10812, General Electric Company, March 1973. 147 013 4.3-15 Rev. 1 1. J. D. LeBlanc, " Maine Yankee Atomic Power Station Startup fest Report," Maine Yankee Atomic Power Company, June 1973. 5. Brookhaven National Labo.'atory interim report files maintained by Core Performance Branch, Task 2 " Moderator Coefficients," and Task 3, " Control Rod Worths." 6. M. K. Drake, ed., " Data Formats and Procedures for the ENDF Neutron Cross Section Library," BN'.-50274 (ENDF-102), National Neutron Cross Section Center, Brookhaven National '.aboratory (1970). 7. Branct. Technical Position CPB 4.3-1, " Westinghouse Constant Axial Offset Cor. trol," l July 1975, attached to SRP Section 4.3. 9 Rev. 1 4.3-16 BRANCH TECHNICAL POSITION CPB 4.3-1 WESTINGHOUSE CONSTANT AXIAL OFFSET CONTROL (CACC) A. BACKGROIND In connection with the staff review of WCAP-bl35 (17x17), we reviewed and accepted a scheme dereloped by Westinghouse for operating reactors that assures that throughout the core cycle including during the most iimiting power maneuvers the total peaking factor, F, will not exceed the value consistent with the LOCA or other limiting accident q analysis. This operating seneme, called constant axial offset control (CAOC), involves maintaining the axial flux difference within a nTrrow tolerance band around a burnup-dependent target in an attempt to minimi'.e the variation of the axial distribution of xenon during plant me euvers. Originally (early '74), the maximum a'lowable Fq (for LOCA) was 2.5 or greater. Later (late '74), when needed changcs were made to the ECCS evaluation model, Westinghouse, in order to meet physics analysis commitments to all its customers at virtually the same time, did a generic analysis (one designed to suit a spectrum of operating and soon-to-be operating reactors) and showed that most plants could meet the requirements of Appendix K and 10 CFR 50.46 (i.e., 2200 F peak clad temperature) if Fq $ 2.32.

Also, I

Westinghouse showed that CAOC procedures employing a + 5% target band would limit peak F q for each of these reactors to less than 2.32. We recognized at that time, however, that not all plants needed to maintain F beiow 0 2.32 to meet FAC, or, needed to operate within a + 5% band to achieve F 1 2.32. In fact, Point Beach was allowed to operate with a wider band because the Wisconsin Electric Power Ccmpany demonstrated to our satisfaction that the reactors could be maneuvered within a wider band (+6,-9%) and still hold F below 2.32. We fully expected that in q time most plants wculd ha/e individual CAOC analyses and procedures tailored to the requirements of their plant specific ECCS analyses. Therefore, wtan we accepted CAOC it was not just F = 2.32 and a + 5% band width we were q approving, but the CAOC methodology. This is analcgous to our review and approval of ECCS and fuel performance evaluation models. The CAOC methodology, which is descrioed in Reference 1, entails (1) establishing an l envelope of allowed power shapes and power densities, (2) devising an operating strategy for the cycle which maximizes plant flexibility (maneuvering) and minimizes axial power shape changes, (3) demonstrating that this strategy will not result in core conditions that violate the envelope of permissible core power characteristics, and (4) demonstrating that this power distribution control scheme can be effectively supervised with excore k detectors. h 1 0 015 4.3-17 Rev. 1 Westinghouse argues that point 3, above, is achieved by calculating all of the load-foll3w maneuvers planned for the proposed cycle and showing that the maximum power densities expected are within limits. These calculations are performed with a radial / axial synthesis method which has been shown to predict conservative power densities when compared to experiment. While we have accepted CAOC on the basis of these analyses, we have also requ "ed that power distributions be measured throughout a number of representa-tive (frecuently, limiting) maneuvers early in cycle life to confirm that peaking factors are no greater t,an predicted. Additionally, we are sponsoring a series of calculations at BNL to check aspects of the Westinghouse analysis. The power distribution meast.rement tests described above will, of course, automatically relate incore and excore detector responses, and thereby validate that power distribution control can be managed with excore detectors. B. BRANCH TECHNICAL POSITION An applicant or licensee proposing CAOC for other than F = 2.32 and al = ;A is expected l q to provide: 1. Analyses of F x cower f raction showing the maximum F (z) at power levels up to 100% q q and DNB performance with allowed axial shapes relative to the design bases for overpower and loss of flow transients. The envelope of these analyses must be shown to be valid for all normal operating modes and anticipated reactor conditions. (See Table 1 of Reference 2 for the cases which must be analyzed to form such an envelope.) 2. A description of the codes used, how cross sections for cycle were determined, and what F values were used. 3. A commitment to perform load-follow tests wherein F is determined by taking incore q maps during the transient. (NOTE: Westinghouse has outlined for both the NRC staff and the ACRS an augmented startup test program designed to confirm experimentally the predicted power shapes. This pr3 gram is presented in a Westinghouse report (Ref. 3). The tests will be carried out at several representative - both 15x15 and 17x17 - reactors. We have endorsed these tests as has the ACRS in its June 12, 1975 letter for the Diablo Ca: yon plant. In addition, for the near term, we plan to require that these licensees who propose to depart from the previously approved peaking factor and target band width perform similar tests, precisely which ones to be determined on a case-by case basis, to broaden our confidence in analytical methods by extending the comparison of prediction with measurement to incluoe more and more burnup histories.) C. REFERENCES 1. T. Morita et al., " Power Distribution Control and Load following Procedures," WCAP-8385 (proprietary) and WCAP-8403 (nonproprietary), Westinghouse Electric l Corporation, September 1974. Rev. 1 4.3-18 2. C. Eiche ger, Westinghouse Electric Corporatien, letter to D. B. Vassallo, U.S. Nuclear..egulatory Commission, July 16, 1975. 3. K. A. Jones et al., " Augmented Startup and Cycle 1 Physics Program," WCAP-8575, Westinghouse Electric Corporation, August 1975. 9 147 017 4.3-19 Rev. .